This invention relates to a novel zirconium alloy, and more particularly to a zirconium alloy having superior corrosion resistance which is suitable as a structural material in a nuclear reactor which material is to be used in contact with water of a high temperature under a high pressure.
A zirconium alloy has an excellent corrosion resistance and a small neutron absorption cross section, so that it is used for producing a fuel assembly channel box 11, a fuel cladding tube 17, or the like which are structural members in an atomic power plant reactor as shown in FIG. 1. There are known, as a zirconium alloy used for these applications, zircalloy-2 (consisting essentially of about 1.5 wt % of Sn, about 0.15 wt % of Fe, about 0.1 wt % of Cr, about 0.05 wt % of Ni, and the balance zirconium), and zircalloy-4 (consisting essentially of about 1.5 wt % of Sn, about 0.2 wt % of Fe, about 0.1 wt % of Cr, and the balance zirconium). In the atomic power plant reactor shown in FIG. 1, reference numeral 10 represents a fuel assembly; 14 a nuclear fuel element; 18 an end plug; 19 an embedded bolt; 20 a space; and 24 a nuclear fuel material supporting means. Even in a zirconium alloy having excellent corrosion resistance, when it is subjected to water or steam at a high temperature and under a high pressure for a long time in the reactor, the oxide coating layer will become thick or the locally concentrated nodule-like corrosion (hereinafter, referred to as "nodular corrosion") will be caused, so that a thickness of non-oxidized portion will be reduced, with the result that the corrosion becomes a factor of decrease in strength of structural members.
To realize large degree of burn-up of atomic power nuclear fuel and to assure long-period operation, it is necessary to further improve corrosion resistance of a conventional zirconium alloy.
For improvement in such corrosion resistance of the zirconium alloy, a method is known for changing the distribution state of intermetallic compound phase (Zr (Fe, Cr).sub.2, Zr (Ni, Fe).sub.2 or Zr.sub.2 (Ni, Fe)) in the metal structure of the zirconium alloy by use of heat treatment. According to a Japanese Laid-Open Patent Publication No. 110412/76, there is disclosed a method of cooling the intermetallic compound phase, which has been evenly dispersed in a crystal grain and at a grain boundary, at a relatively slow cooling rate (30.degree.-200.degree. C./s) from a range of [.alpha.+.beta.] phase coexisting temperature. According to a Japanese Laid-Open Patent Publication No. 70917/77, a method is disclosed which includes the steps of: quenching the zirconium alloy (at a cooling rate .gtoreq.800.degree. C./s) from a temperature range, at which a single phase of .beta. occurs, to provide solid-solution in which alloying elements constituting intermetallic compound phase are substantially completely in solid-solution; and annealing the zirconium alloy in a temperature range, at which .alpha. phase occurs, to selectively precipitate intermetallic compound phase at grain boundaries.
However, the precipitation of Fe, Cr or Ni at the grain boundary as intermetallic compound phase by use of these methods causes the amount of Fe, Cr or Ni existing in solid-solution of crystal grains to be reduced, resulting in deterioration of corrosion resistance of crystal grain. The inventors have discovered that such decrease of Fe, Cr or Ni in solid-solution is apt to cause the nodular corrosion progressing from a point within the crystal grain.